1. Field of the Invention
This invention relates to a method and apparatus for monitoring and analyzing on an on-line basis the integrity of fluid containing vessels, and particularly nuclear reactor vessels, during both normal and abnormal fluid temperature and pressure transients. More particularly, the invention is directed to presenting to a plant operator a visual representation of the margin between the reference nil-ductility transition temperatures required for crack initiation and arrest, and the actual reference nil-ductility transition temperature through the entire thickness of the vessel at selected critical points taking into account, in the case of reactor vessels, the effects of radiation embrittlement. The invention also encompasses techniques for providing accurate determinations of reactor coolant temperature at critical locations in a nuclear reactor pressure vessel, even during injections for core cooling purposes of cold coolant water under stagnant flow conditions.
2. Background Information
The pressure vessel in which the core of a nuclear reactor is housed is subjected throughout its nominal 40 year life to stresses induced by changes in reactor coolant temperature and pressure. During normal operations, these stresses are of moderate rate and amplitude and have no significant adverse affect on the vessel which is designed with a large safety margin to withstand the expected loading. These normal operations include heat up and cool down where restrictive schedules are rigorously followed to maintain the stresses within prescribed limits.
During some abnormal operations, such as a loss of coolant accident (a LOCA), temperature and/or pressure transients which far exceed those which occur during normal operations may be experienced. Though some abnormal events may not impose serious stresses on the vessel themselves, corrective action taken to alleviate the initial problem, such as injecting cold water into the reactor, may lead to transients which could be a threat to vessel integrity.
The reactor pressure vessel is a cylindrical enclosure with hemispherical ends which is fabricated from steel plates welded along longitudinal and circumferential seams. The upper hemispherical end, or head, is removable for access to the internals. Inlet and outlet nozzles, for each reactor coolant loop, typically 2 to 4 in a pressurized water reactor, are welded into the vessel walls. Typically, it is the welds where flaws are likely to be found which can develop into cracks under the stress induced by large thermal transients. The problem is compounded, in general, by embrittlement of the metal by the neutron radiation to which the vessel is subjected and is of most concern with regard to the welds at the level of the reactor core where the fluence is the highest.
It is necessary for the operator of a nuclear power plant to be constantly aware of the status of the reactor pressure vessel with respect to non-ductile failure. Presently, the operator has two means of obtaining this status: standard heatup/cooldown curves and the recently developed status tree approach. The heatup/cooldown curves define the allowable pressure and temperature domain and are mandated by the Nuclear Regulatory Commission for use during normal startup and shutdown of the reactor. These curves are generated in accordance with Appendix G of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (A.S.M.E. code). They can also be utilized during an abnormal event to determine if the temperature and pressure transients remain within the limits established by the curves. The status tree approach, which was developed expressly to assess vessel integrity during pressurized thermal shock events, assumes a step change in temperature to provide a worst case scenerio and through the application of off-line fracture mechanics analysis establishes pressure and temperature boundaries for a particular plant and provides instructions for actions to be taken as the boundaries are approached or exceeded.
Both of these presently available methods of determining vessel integrity status are limited in the following ways: (1) they utilize a quasi-time dependent approach to determining vessel status, when in fact the fracture mechanics problem is a very time dependent function of both vessel temperature and pressure, which requires the introduction of excessive, unquantifiable conservatism into the status information since the operator is not provided with accurate knowledge of the actual margin to flaw initiation in the vessel wall as the plant is stabilized and recovered from a cooldown transient; (2) both methods use cold leg resistance temperature detector (RTD) temperatures as the indicator of current reactor vessel temperature thereby introducing uncertainties into the measurements, and rendering the determined status potentially imprecise, particularly in the event of thermal stratification as a result of safety injection flow under stagnant loop conditions; and (3) both methods require operator interpretation of the cooldown history in order to provide a status, which is a time consuming and subjective task that must be performed, at times, under accident conditions.
Under present practice, the NRC requires that if the heatup/cooldown curve limits are exceeded during a thermal transient, an analysis must be performed before the conditions have occurred which could cause potential flaws to be initiated in the vessel. Such an analysis is performed after the fact and can cause a delay in returning the unit to power where the results indicate that critical conditions were not reached. Of course, if the analysis indicates that critical conditions had existed, then detailed inspections and/or repairs may have to be performed.
In the post event analysis of an abnormal plant transient, data from the event are used to construct temperature and the resulting stress profiles through the vessel at selected critical locations. Through the application of fracture mechanics analysis the stress profile is used to calculate how close the vessel may have come to non-ductile failure. Specifically, the procedure postulates flaws of varying depth at the critical location and calculates for each such flaw the margin between the stress intensity factor and both the fracture initiation toughness and arrest toughness of the material calculated as a function of the actual reference nil-ductility transition temperature RT.sub.ndt, which in turn, is a function of the condition of the material and the fluence. If these margins for all of the postulated flaws meet preset limits, it is assumed that no damage was caused by the transient. Application of such post event fracture mechanics analysis of thermal shock events is discussed in a paper entitled "Method for Fracture Mechanics Analysis of Nuclear Reactor Vessels Under Severe Thermal Transients", by C. B. Buchalet and W. H. Bamford and published by the American Society of Mechanical Engineers, 1975.
The results of such after the fact facture mechanics analysis have been cross-checked at times by required RT.sub.ndt analysis. In this analysis, the required RT.sub.ndt for crack initiation is determined by finding what the RT.sub.ndt would have to be to make the fracture toughness for crack initiation equal to the stress intensity factor for various depth cracks through the vessel wall. The required RT.sub.ndt for crack arrest is determined in a similar manner using the fracture toughness for arrest. Plots of the required RT.sub.ndt for crack initiation and arrest are overlayed with a plot of the actual RT.sub.ndt to determine if there are any locations where the actual RT.sub.ndt exceeds the required RT.sub.ndt and at what depths such crack propagation would arrest. As mentioned, such required RT.sub.ndt analysis has been used previously as a check on the after the fact analysis performed in accordance with the techniques set forth in the Buchalet et al paper. It has also been used as an analytical tool to determine what the effect would be of raising the temperature of the normally ambient temperature injection water.
In any analysis of vessel integrity, the actual transient to which the vessel is exposed is crucial. The pressure transient is readily available; however, the temperature transient at the critical locations such as at welds in the vessel beltline, is much more difficult to determine, especially where cold safety injection water is introduced under stagnant flow conditions. Since it is not practical to place temperature sensors at such locations, other means must be utilized. As mentioned above, a conservative approach is to assume a step change to the temperature of the safety injection water, but this can lead to the plant operator addressing vessel integrity, when it is not important, and the operator should be addressing other more important critical safety functions at that time. A more recent approach is the mixing cup analysis which utilizes data generated from one-fifth scale vessel model tests conducted at Creare Research Laboratories. This analysis utilizes a mathematical model of a selected volume of the reactor coolant system to analytically determine the mixed mean temperature at the core midplane in the downcomer. While this method has been successfully used in after the fact analysis of events, it has not been adapted to on-line determination of beltline temperature due primarily to the difficulty in defining the boundary conditions in real time.
It is a primary object of the present invention to provide an accurate, easily understood, on-line, real-time representation of the integrity status of a vessel containing a fluid subject to varying temperature and pressure conditions and especially the pressure vessel of a nuclear reactor.
It is also an object of the invention to achieve the above through improved accuracy in determining the temperature of the reactor coolant adjacent the critical locations in the pressure vessel even during safety injection under stagnant flow conditions.
It is another object of the invention to realize the first objective through generation of a display which presents the operator with a real-time visual indication of the current margin to non-ductile failure for selected critical pressure vessel locations.